The invention relates to heat exchangers and more particularly to heat exchangers with auxiliary cooling systems.
In nuclear steam supply systems well known in the art, a reactor vessel contains fuel assemblies comprising fuel elements with nuclear fuel therein which produce heat in a commonly understood fashion. A primary coolant, which in fast breeder reactors may be liquid sodium, is circulated through the reactor vessel in heat transfer relationship with the fuel assemblies therein transferring heat from the fuel assemblies to the primary coolant. The primary coolant may then be conducted by a piping network to a heat exchanger and back to the reactor vessel tracing a path that is generally referred to as a primary loop. While passing through the heat exchanger, the primary coolant transfers heat to a secondary coolant which may be liquid sodium. The secondary coolant may then be conducted to a steam generator that produces steam in a manner well known to those skilled in the art. The path traced by such a secondary coolant is generally referred to as a secondary loop. The secondary loop is interposed between the primary loop and the steam production phase in order to isolate the primary coolant which is radioactive from the steam production phase. In many commonly known nuclear steam supply systems, there are three primary loops disposed symmetrically with respect to the reactor vessel each having a coolant pump disposed therein to pump the primary coolant through each primary loop.
During reactor operation, the three coolant pumps simultaneously pump primary coolant into the reactor vessel where the three primary coolant streams intermingle and pass in heat transfer relationship with the fuel assemblies. From this common pool of primary coolant, the primary coolant exits the reactor vessel, flows through the piping network, through the heat exchangers, where heat is transferred from the primary coolant to the secondary coolant, and back to the reactor vessel thus cooling the nuclear reactor. It is now well understood that a nuclear reactor generates a tremendous amount of heat. While the coolant pumps are capable of pumping primary coolant through the reactor vessel in sufficient volume to cool the reactor, auxiliary cooling systems are provided to cool the reactor should the normal heat removal systems not be operable.
One such auxiliary cooling system is one used in conjunction with the heat exchanger disposed between the primary and secondary loops of a liquid metal cooled fast breeder reactor. In that concept, a cylindrical housing defines within it, an inner annulus wherein vertical heat transfer tubes are diposed. An outer housing disposed around the cylindrical housing defines therebetween an inlet annulus that is in fluid communication with the inner annulus. An auxiliary cooling coil is disposed in the inlet annulus and carries a third coolant. A coolant pump pumps primary coolant into the inlet annulus where it flows upwardly across the auxiliary cooling coil and enters the inner annulus through openings near the top of the heat exchanger. After entering the inner annulus, the primary coolant flows downwardly through the inner annulus and in heat transfer relationship with the heat transfer tubes in the inner annulus. In this manner, heat is transferred from the primary coolant to the secondary coolant thus providing a mechanism for cooling the reactor. In this concept should the coolant pump that pumps the primary coolant fail, an auxiliary pump is activated to pump the primary coolant. However, this auxiliary pump is small or itself may not be operable which may thereby limit the flow of primary coolant. In this prior art concept should, for whatever reason, the primary or secondary coolant flow become minimal a third pump is activiated that pumps a cool third coolant through the auxiliary cooling coil. The primary coolant in the inlet annulus flowing upwardly across the auxiliary cooling coil thereby transfers heat to the third coolant. This concept thus provides an auxiliary system for cooling the primary coolant which in turn cools the nuclear reactor. There is, however, a disadvantage to this concept. When the primary coolant in contact with the cooling coil in the inlet annulus is cooled, it may begin to flow downwardly in a reverse manner in the inlet annulus due to the natural circulatory effect. This downward-reverse flow of primary coolant has the effect to limit the flow of primary coolant through the primary loop. This presents a particularly acute problem, because the auxiliary cooling system is intended to be utilized when there is a minimal flow of primary coolant through the primary loop. Thus the effectiveness of such an auxiliary cooling system may be limited by this reverse flow characteristic.
In U.S. Pat. No. 1,311,336 to J. O. Frazier, issued July 29, 1919, there is described a by-pass flue gas cane juice heater wherein hot flue gases are diverted by a damper through a by-pass duct passing in heat transfer relationship with tubes carrying cane juice, thereby heating the cane juice. Also, the damper may be automatically closed to avoid overheating the cane juice. While the Frazier patent describes one arrangement for avoiding overheating of cane juice, it does so by mechanically not diverting and thus minimizing flow rather than by increasing flow of a coolant by means of natural circulation.